Material Testing Reactor Research Articles

Study of the fuel irradiation capsule using NEPTUNE_CFD software

The Fuel Irradiation Capsule (FUICA) is an experimental device designed to irradiate fuel elements in a Material Testing Reactor (MTR) during long period. This component is being studied for implementation in the Jules Horowitz reactor (RJH) under construction in France (1).Its function is to irradiate and to cool fuel sample on a passive way by transferring the thermal energy generated to the device periphery, through an annular zone of confined and pressure-regulated water (2). Its external wall, kept cold by the reactor cooling circuit, allows for the heat extraction. Natural convection is the driving force for the heat transfer between the fuel element and the cold wall of the device. By eliminating any internal forced convection cooling system, the device becomes relatively simple and cost-effective.This paper describes the device modeling carried out using the NEPTUNE_CFD software (3). The present work, in its first part, focuses on the behavior of the device depending on the heat flux, and in its second part, the sensitivity to boundary conditions. The objective is to provide a way to design and optimize the fuel capsule thermal characteristics. For that, an arbitrary size of fuel capsule is chosen in order to analyze the physics involved. A sensitivity study is then presented in order to highlight the influence of the physical parameters on the fuel capsule performance.Three operating ranges for the heat transfer are exhibited in this study: moderate in the liquid phase, the heat transfer becomes optimal in the two-phase regime, and even improves with the level of heat flux before deteriorating abruptly for high flux called Critical Heat Flux (CHF). Two observations also arise to improve thermal transfer: in the liquid phase, it is necessary to reduce the water thickness to improve convection, and in the two-phase regime, the heat transfer increases proportionally to the heat flux. In the end, the irradiation capsule is capable of extracting very high heat flux.

Fuzzy reliability algorithm for the shutdown system of research reactor

Abstract Probabilistic safety assessment (PSA) has been applied to evaluate the safety of nuclear reactor systems. The results of PSA are used by designers, utilities and regulatory body to confirm the reactor design, to change operation procedures, to enhance the reliability of safety systems, or to modify regulation. Fault tree analysis (FTA) has been used as a deductive means for PSA. System components are assumed to always have defined failure probabilities; however, in reality this assumption is questionable and sometimes there is insufficient data about the failure probability of some components. To solve this problem, fuzzy approaches have been proposed and implemented. The aim of this study is to apply a fuzzy reliability algorithm to estimate basic events failure probabilities for the shutdown system of a research reactor. The reactor is a 22 MW light water open pool type material test reactor and the shutdown system is responsible for the fast shutdown of the reactor whenever safety limits are exceeded. Then, to illustrate the ability of the algorithm, the results are compared with real basic event failure probabilities. The results verify the competence of the algorithm and confirm that it can be applied as an alternative method when quantitative data are not available. Also, a risk importance measure is done on the fault tree of the shutdown system to illustrate which components need to be focused for improvements to achieve the safe operation of shutdown system.

Evaluation of U3O8–Al, UO2–Zr, and U3Si2–Al fuel plates in a hypothetical melting accident of Tehran research reactor (TRR)

Tehran research reactor (TRR) is a material test reactor (MTR) that uses U3O8–Al fuel plates. Fuel plates are usually used in research reactors due to their great thermal characteristics. The occurrence of various accidents in nuclear reactors has increased the efforts to develop accident tolerant fuels (ATF). For this reason, intermetallic compounds of uranium and silicon, have been introduced as candidates for ATF. In this study, the heat transfer and melting of fuel plates composed of U3O8–Al, UO2–Zr, and U3Si2–Al, in normal operation and hypothetical heat transfer impairment in the TRR have been studied. These compounds are chosen as candidates for probable replacement in the new design of the TRR. In this work, two-dimensional equations of heat transfer and enthalpy using the finite difference method have been solved and solid to liquid phase changes are discussed. According to the findings of this research, during a hypothetical heat transfer impairment, UO2–Zr fuel undergoes a phase change from solid to liquid after a longer period of time compared to the other two fuels. This delay is attributed to the relatively high melting point of UO2–Zr, which is notably higher than the melting points of the other two fuels. It has been shown that when the focus is on transferring heat from the fuel plate to the coolant, U3Si2–Al performs better than the other two types of fuel due to its higher thermal conductivity and relatively good melting behavior.

Effect of Cooling Pump Transient Power Supply on ETRR-2 Safety Parameters

The primary goals of the engineering design for nuclear reactors involve safeguarding the integrity of the reactor core. Preserving the integrity of the cladding material is especially crucial, as it serves as the initial defense against the potential dangers posed by radioactive materials. In this work, an accident analysis of core cooling pump power transients of different ratios of the nominal pump power in a typical material test reactor is conducted. Phase failure is a very common electrical fault experienced by three-phase motors. Pump power reduction can be initiated due to several causes, like phase failure, voltage reduction, winding failure, and other causes. The nuclear reactor analysis code PARET/ANL version 7.6 is used to carry out these calculations. The accident scenario began with the reactor operating steadily, then experiencing a transient in the core pump power. This caused the core flow rate to decrease and eventually stabilize at a lower level as the pump power decreased. Core cooling pump power variations ratios of 20%, 33.3%, 50%, and 70% of the nominal pump power are considered in this work. The accident analysis is conducted under the availability and unavailability of reactor safety systems. Reactor safety parameters are reported for all cases of the core pump power variations.

The effect of microstructure on recovery and recrystallization after annealing of two neutron irradiated ITER specification tungsten

The study of the recrystallization resistance of engineering grade tungsten (W) prior to and after neutron irradiation as well as subsequent thermal annealing provides valuable insight into neutron induced defect interactions and their kinetics as well as their correlation with recrystallization mechanisms. The outcome of such investigations would be of assistance for appropriate grade selection for a fusion reactor and the design of healing processes enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification W, forged bar and cold-rolled plate, were neutron irradiated to a dose of 0.18 dpa at 600 °C in the Material Test Reactor BR2, Mol, Belgium, and subsequently isochronally annealed from 700 to 1550 °C for 24 h with a 50 °C step with two non–irradiated counterparts. X-ray diffraction, optical microscopy and positron annihilation lifetime spectroscopy, combined with Vickers Hardness are used to correlate the recrystallization process and recovery mechanisms. It is found that the non-irradiated plate has a higher recrystallization resistance compared to that of the bar, with the former requiring 100 °C higher temperature for 24 h annealing time to recrystallize. Despite the irradiation induced dislocation loops and voids being of similar sizes and densities for both grades, the different initial microstructure affects the recrystallization process upon annealing, delaying it by 50 °C on the plate. Further in the plate a complete void dissolution is observed before recrystallization, which is not the case for the bar, for which the two effects occur at the same temperature.

JENDL-5.0 nuclear data sensitivity, uncertainty, and similarity analyses on the criticality of RSG GAS multipurpose research reactor

The criticality, sensitivity, uncertainty and similarity analysis of the clean, first core of the G. A. Siwabessy Multipurpose Reactor (RSG GAS), using the newly released Japanese Nuclear Data Library version 5 (JENDL-5) was conducted to contribute the validation effort of the JENDL-5, especially for the application on beryllium reflected, light-water moderated, low-enriched uranium (19.75 wt%) fueled material testing reactors. The JENDL-5 sensitivity coefficients and covariance matrices were prepared by MCNP6 and AMPX-6, respectively. The keff uncertainty was then evaluated with TSUNAMI-IP module of SCALE-6. The same module was used for similarity analysis by comparing the sensitivity data of 3,690 criticality experiments. The criticality (keff) analysis results ([C/E-1]) showed maximum overestimation of 801 pcm, 477 pcm and 547 pcm, for JENDL-5, JENDL-4.0 and ENDF/B-VIII.0, respectively. The keff uncertainty originated from the nuclear data was estimated to be 620 pcm, 644 pcm and 637 pcm for JENDL-5 only, JENDL-5 & ENDF/B-VIII.0, and ENDF/B-VIII.0 covariance data, respectively, which are comparable with the keff ([C/E-1]) values. About 125 experiment cases were found to have correlation factors of > 0.8, however, no experiment showed a strong correlation factor (>0.9), therefore, in the future criticality experiments similar to the RSG GAS should be conducted.

Validation of the dynamic simulation capabilities of Serpent2/Subchanflow using experimental data from the research reactor SPERT IV D-12/25

Research nuclear reactors are critical to the development of nuclear technology, but because of the complex configuration of the fuel assemblies and the different initial operating conditions, neutronic and thermal–hydraulic analyzes are performed with dedicated or adapted codes. These simulation tools might not be as accurate, since they use simple resolution methods and assumptions that do not always capture all aspects of the behavior of a nuclear research reactor. On the other hand, as time goes by, the evolution of numerical tools for the core analysis of power reactors have experienced a considerable progress. Nowadays, pin/subchannel level analysis of the core with coupled codes based on transport (SP3, MOC, SN, etc.) or Monte Carlo methods are applied in addition to the nodal diffusion codes. Hence, the research community is adapting and validating selected high-fidelity tools developed for power reactors to perform detail core analysis of e.g. Material Testing Reactors (MTR) cores at plate and subchannel level.This work deals with the validation of the high-fidelity coupled Serpent2/Subchanflow, which was modified and extended for the plate/subchannel analysis of MTR-cores, using the data of rod ejection tests performed in the SPERT IV D-12/25 reactor, especially the tests B-34 and B-35 were selected to validate the dynamic capability of Serpent2/Subchanflow. In these unique tests, experimental data e.g. thermal neutron flux, core power evolution during the rod ejection tests, and the plate cladding temperature was measured. It is noted that, due to the lack of detailed information on the initial conditions, the extraction and introduction scenarios of the transient rod for the reactivity insertion required calibrations and assumptions regarding velocity and position.The comparison of selected parameters predicted by the coupled simulations at plate/subchannel level of the SPERT IV reactor with the measured data at static and transient conditions shows excellent agreement confirming the high accuracy appropriateness of the used code for the analysis of research reactors. The calculated values of thermal neutron flux and core power evolution have a statistical error of ± 2 sigma. It was also found that the maximum temperature difference between calculated and experimental values is 7 °C and ∼ 10 °C for tests B-34 and B-35, respectively. In addition, the coupled code predicts for the first time the temperature of each plate and subchannel considering the local feedbacks between neutronics and thermal-hydraulics allowing the identification of the hottest/coldest plate in the core. The high-fidelity validation tool can provide comparison solutions for current research reactor core analysis methods, such as core analysis with point kinetics or 3D nodal diffusion codes coupled with fuel assembly-level thermal-hydraulics.

High-fidelity models for the online control of power ramps in fuel displacement systems – Revisiting the ISABELLE experiment in the OSIRIS Material Testing Reactor

The Jules Horowitz Reactor (JHR) is an MTR under construction at the CEA Cadarache, France. Its design and future operation build upon the lessons learned from the OSIRIS MTR. The CEA intends to transfer the knowledge accumulated with the ISABELLE1 loop of OSIRIS to the ADELINE loop of the JHR, both dedicated to power-ramp-type irradiation experiments, the purpose of which is to test the resistance of PWR fuel rod cladding under extreme levels of stress. In this article, we revisit the ETALISA experiment, a heat balance measurement experiment performed in 1992 for calibrating power ramps in ISABELLE1. We use high-fidelity modelling and simulation tools, especially the neutron-gamma TRIPOLI-4® Monte Carlo transport code, to calculate the detailed components of the heat balance, correction terms, and uncertainties. Comparisons between the simulations and the experiments show a very good agreement in the total linear heat generation rate of 400 W/cm at high power. The computed 2σ uncertainty is found to be 5%, a value essentially identical to the estimate derived in 1993 from an engineering approach. The use of modern simulation tools does not make it possible to improve upon this value, but provides a better understanding of the various components and corrections introduced in the total heat balance. The main limitations come from the ISABELLE1 online instrumentation, thermocouples and self-power neutron detectors, which set a limit on our very knowledge of the actual power ramp experimental conditions.

A standard capsule design for structural material testing in the Advanced Test Reactor

Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x1019 to 2.3x1021 n/cm2, depending on experiment position. This paper presents and discusses the standard capsule design and analysis.

Multi-parameter optimization of gamma emission tomography instruments for irradiated nuclear fuel examination

Material test reactors have an extended use in irradiation testing of novel nuclear fuel materials and the fuel behavior in off-normal conditions. The performance of the nuclear fuel is examined in in-pile and out-of-pile post-irradiation examinations (PIEs), e.g., using Gamma Emission Tomography (GET). GET is a nondestructive assay that images the internal spatial distribution of gamma-emitting nuclides built up in the fuel due to irradiation. Since GET can be performed close to the reactor and without intrusion in the fuel object, it can potentially speed up the data generation from PIE in irradiation testing.The performance metrics of GET devices can be identified regarding time requirements, noise in the reconstructed image, signal-to-background ratio, and spatial resolution. However, these are complicated to determine, partly due to inherent trade-offs between the metrics themselves, partly because they depend on the fuel activity and its spectrum (i.e., object dependent), and, finally, on the GET setup and its configuration.This work proposes a structured methodology for optimizing the collimator design for a new generation of GET tomography setups, intending to improve spatial resolution by one order of magnitude: from the millimeter scale to the hundred-micron scale. The conflicting performance metrics are determined based on the controllable parameters of the GET setup and the uncontrollable parameters of an anticipated fuel object, able to provide a signal-to-background ratio above a customized limit, chosen based on the specific application. The trade-off between the performance remaining metrics is then visualized by a Pareto approach, where dominated solutions are rejected. Finally, constraints on noise level and measurement time are used to find the optimal spatial resolution, without applying noise suppression filters.Two GET setups are presented using the outlined method. Firstly, to upgrade the tomography test bench BETTAN at Uppsala University, a new segmented HPGe detector is planned to be used with low-activity fuel rod mock-ups. Secondly, a GET system for investigating high-activity nuclear fuel rods of representative burnup. For a nuclear fuel inspection, the results showed that a spatial resolution of about 300 μm is possible with reasonable noise and measurement time constraints.

Heat transfer characteristics for multi-silicon ingots irradiation in a typical research reactor

A computational fluid dynamics (CFD) study is conducted for silicon ingot irradiation facility in a typical material testing reactor (MTR). The ingots are placed inside the irradiation device in staggered and in-line bundles. The configurations are varied according to the variations of longitudinal and transverse pitches Sl and St, respectively. For in-line bundles, Sl and St are changed in the range of 1.1, 1.2, 1.3, 1.4, and 1.5 of the ingot diameter. For staggered bundles, Sl is changed in the range of 1.1, 1.2, 1.3, 1.4, and 1.5 of the ingot diameter while St is changed in the range of 0.95, 1.0, 1.05, 1.1, and 1.15 of the ingot diameter. The silicon ingot volumetric heat generation Qv is allowed to change around its nominal value in the range of 0.25, 0.5, 0.75, 1.0, and 1.25 W/g. The effects of rotation of ingots during irradiation at velocities of 15, 30, and 60 rpm are investigated. The computational fluid dynamics code ANSYS FLUENT is employed to carry out the calculations. The temperature profile inside the silicon ingot during irradiation is of significant importance for uniform doping. Therefore, the impact of heat generation and bundle pitches on the temperature profile through the ingot is calculated and reported for all cases of this study. The results show that the maximum ingot temperature does not exceed the limits for each bundle at maximum heat generation. Two correlations for average Nusselt number for both aligned and staggered configurations are generated to describe heat transfer rates during natural convection regime. The rotation of ingots during irradiation showed considerable enhancement of heat transfer rates.

Effect of neutron irradiation on tensile properties of advanced Cu-based alloys and composites developed for fusion applications

The effect of neutron irradiation on tensile properties and fracture mode has been investigated for several advanced CuCrZr alloys in the frame of the European fusion material development program. Five material grades utilizing different strengthening principles have been exposed to neutron irradiation up to ∼2.5 dpa (displacement per atom) in the target operational temperature range of 150–450 °C. The strengthening mechanisms are based on the application of: i) tungsten particles; ii) tungsten foils (laminate structure); iii) tungsten fibers; iv) Y2O3 particles; v) vanadium addition (0.22%). Neutron irradiation was performed in the BR2 material test reactor inside the fuel channel in order to maximize the fast neutron flux. The upper irradiation temperature of 450 °C was selected to validate the ability of the pre-selected advanced grades to sustain the high temperature irradiation, since the baseline ITER specification CuCrZr is known not to retain sufficient tensile strength above 400 °C in non-irradiated conditions and shows strong irradiation induced softening above 300 °C. Neutron irradiation at 150 °C caused severe embrittlement of tungsten-copper laminates as well as a considerable reduction of the total elongation of all other grades. The irradiation at 450 °C led to the reduction of the yield strength and ultimate tensile strength (i.e. irradiation softening) in the vanadium-doped alloy similar to CuCrZr, while all other materials preserved or increased their strength (irradiation hardening). The fracture surfaces of the tested samples were analysed to investigate the modification of the deformation mechanisms in each particular case.

Numerical Simulations of Flow-Induced Deflections in MITR LEU Fuel Plate Due to Channel Size Disparity

The hydromechanical stability of the fuel plates in parallel coolant channels of a Materials Testing Reactor (MTR) fuel element design is of great importance to the safety of research and test reactors. Previous analytical, experimental, and numerical efforts focused on parallel channels with the same or similar size; also, in the prior numerical simulations, the fuel plate was often assumed to be perfectly flat. This work presents the results of a fluid-structure interaction simulation performed to evaluate the flow-induced deflections of the fuel plates in the low-enriched uranium (LEU, <20 wt% 235U) fuel element design for the conversion (from highly enriched uranium) of the Massachusetts Institute of Technology Reactor (MITR-II, also referred to as MITR). Various manufacturing and assembly tolerances of the MITR LEU elements are considered in the analysis, and the effects of channel size disparity, nonideal plate shape, and flow rate uncertainty are investigated. Results show that, for all cases analyzed, the deflection occurs toward the larger channel, and the change in any channel stripe remains small (less than 0.021 mm) compared to fabrication tolerances. In addition to simulation work, a hydraulic performance test of the MITR LEU fuel element is currently planned to support conversion to the use of LEU fuel.

Fusion-dedicated material irradiation facilities at MYRRHA: Conceptual design and damage equivalence studies

The Belgian Nuclear Research Centre is currently developing MYRRHA - a subcritical accelerator driven system ignited by a 600 MeV superconducting linear proton accelerator, cooled with an eutectic lead bismuth. The implementation of MYRRHA will take place in phases, where phase I, covering 2019–2027, includes the deployment of the 100 MeV proton accelerator and beam-added research facilities. Development of the dedicated facilities to enable irradiation of fusion materials and components is in the portfolio of this R&D block. The baseline concept of the 100 MeV fusion-dedicated facilities has been presented for the first time in 2019 and in this contribution, we extend its description and provide main elements of the conceptual design of the irradiation installations, preliminary configuration of the hot cell facilities (required to load/unload the irradiation modules), information on the irradiation envelope. In addition, we summarize the results of the computational study performed to investigate the equivalence (in terms of damage rate, nuclear transmutation, temperatures and expected irradiation-induced microstructure) of the irradiation conditions at MYRRHA facilities in comparison with other material test facilities such as IFMIF and ESS, and in DEMO as well as in the mixed-spectrum material test reactors. The aspects related to miniaturization of the sample geometries are also discussed to substantiate the feasibility and relevance of this facility for fusion material's R&D programme.

Neutron-Physical and Radiation Characteristics of Different Low Enrichment Fuels (Thorium and Uranium) in a Proposed JRTR Research Reactor

Abstract: The Jordan Research and Training Reactor (JRTR) is a 5-MWth, open-pool type, light-water moderated and cooled reactor with a heavy-water reflector system. The core consists of standard Material Testing Reactor (MTR), the plate type of fuel assemblies with low enriched fuel of 19.75% U-235 enrichment. One of the most important purposes of JRTR is producing several radioisotopes e.g. (99Mo/99mTc, 131I and 192Ir isotopes) by using the neutron activation technology. A new computational model has been developed for different low enrichment fuels (Thorium and Uranium) in the JRTR by using the Serpent Monte Carlo code. The purpose of this paper is to validate Th-U233 fuel by comparison of the calculation results of important neutronics parameters, like keff, flux distribution, kinetics parameters, power peaking factors, heat decay and activity of spent fuel. Keywords: U3Si2, Th3Si2, JRTR, Light-water Reactors (LWRs), Extended cycle lengths.

Investigation of the Fuel Shape Impact on the MTR Reactor Parameters Using the OpenMC Code

The goal of this study was to evaluate the impact of simulating different fuel shapes for the material testing reactor (MTR). Two OpenMC codes were built, and the first OpenMC model was simulated using a curved shape fuel element to mimic the real dimensions and shape of the MTR. The code core parameters were validated with the collected parameters from the experimental work and two well-known Monte Carlo simulation codes (MCNP and SCALE). The validation process included the axial flux profile and criticality. After the OpenMC curve fuel model was validated, the MTR fuel was simulated as flat fuel elements with the exact amount of fuel as in the curve fuel model. By comparing the two OpenMC models’ calculations, it was observed that the radial flux distribution has only a slight difference due to fuel mass similarity. In conclusion, simulating the MTR fuel as flat elements provided a good agreement calculation compared to the real shape, but it was also observed that this might carry some discrepancies for in-depth simulation studies.

Optimization of heat removal and radiation protection of dry cask storage for MTR spent nuclear fuel

To provide a solution for the limited storage space in wet-type storage in Indonesia's material testing reactor (MTR) spent nuclear fuel (SNF) storage, a study on dry cask storage for MTR SNF is needed. This study designed, simulated, and experimentally calculated the safety parameters and the manufacturing cost of dry cask storage simultaneously. The optimization results were validated and analyzed using experimental data from a dry cask storage prototype. Decision variables and constraints were provided as inputs into the MATLAB software to obtain three optimized conditions: safety, cost, and multi-objective optimized conditions. The multi-objective optimization results demonstrated that the thickness of dry storage design concrete and lead (Pb) are 0.06 and 0.51 m, respectively. The height and width of the vent are 0.15 and 0.5 m, respectively, and the vent elevation difference is 2.43 m. For five variables, the optimum value for the canister surface temperature is 66.8 °C and the cost required to manufacture this dry storage is USD 147,827. Moreover, MicroShield validation revealed that the dry cask storage surface exposure is 104.8 mR/h, which is lower than the safe limit of the maximum value specified as 160 mR/h. The simulation of the canister surface temperature using ANSYS Fluent showed that the surface temperature value is close to the calculations performed in MATLAB. Temperature validation of experimental data from the dry storage prototype indicated that the temperature obtained via theoretical and experimental calculations is relatively close: 45.2 and 51.3 °C. This research can be developed to other types of dry storage design by changing the value of the decision variables that is appropriate for the dry storage type to be designed.

Thermal-Hydraulic Analysis of Inward and Outward Fuel Plate Buckling in a Typical MTR Reactor

In the present work, a numerical study of inward and outward buckling of two successive fuel plates of a typical material testing reactor is investigated using computational fluid dynamics code. Fuel plate buckling results in partial blockage of the hot channel. Both buckling toward the inside and outside are considered. Simulations are conducted for different blockage levels of the nominal flow area, i.e., 0%, 20%, 40%, 50%, 60%, and 70% for inward buckling. Blockage levels of 0%, 20%, 40%, 50%, 60%, 70%, 80%, and 90% are considered for outward buckling. The impact of the flow field redistribution in four successive channels on the cooling capacity of each channel is investigated. The obtained results show that for an inward buckling ratio greater than 50%, critical phenomena will occur that could affect the clad integrity. Moreover, for inward buckling of 70%, the maximum clad temperature in the blocked channel reaches the value associated with the onset of nucleate boiling at the operating pressure. On the other hand, for outward buckling of 90%, critical phenomena that could affect the clad integrity will occur.

Defect evolution of neutron irradiated ITER grade tungsten after annealing

The microstructural evolution of neutron irradiated tungsten (W) after annealing and its correlation with the corresponding mechanical properties provides valuable insight on the defect interactions and their annihilation processes. This would result in the identification of recovery mechanisms, leading to the design of healing processes and thus, enabling the lifetime extension of fusion reactor components. Within this framework, samples from ITER specification forged W bar were neutron irradiated to a dose of 0.2 displacements per atom (dpa) at 600 °C in the Belgian Material Test Reactor (BR2) and subsequently annealed at 800 and 1000 °C. The evolution of the irradiation induced defects after annealing has been assessed by transmission electron microscopy (TEM), positron annihilation lifetime spectroscopy and electrical resistivity and its effect on the mechanical properties are discussed in terms of Vickers hardness. Neutron irradiation results in the formation of dislocation loops and voids. TEM observations show an increase in the size of both defect types after post irradiation (PI) annealing at both temperatures, accompanied by an initial increase in their density after PI annealing at 800 °C, followed by a subsequent decrease after PI annealing at 1000 °C. The presence of TEM-irresolvable defects of both types is revealed in the as-irradiated state, which is evidenced by the evolution of Vickers hardness, resistivity and positron lifetimes and their correlation after the PI annealing. In the as-irradiated state, as well as after PI annealing at both temperatures, the hardening is dominated by voids.

Experimental study of different innovative measurement methodologies applied to a canned pulsed eddy current testing probe suitable for dilation measurement of test specimens

A compact pulsed eddy current testing probe is being developed for the online dilation measurement of test specimens, examined in the high-temperature environment of material testing reactors. The probe has to be canned with metallic clad to protect it from the corrosive sodium-potassium coolant medium present in the material testing reactor. Electromagnetic modelling of the probe was carried out to compute the distribution of time-dependent eddy current density in the vicinity of the probe. It is understood that the slope of the pulsed eddy current signals in the specific time zone where the lift-off point of intersection occurs show a good correlation to the distance between the face of the probe and the test specimen. This paper discusses the experimental study of employing different measurement methodologies at room temperature with the canned pulsed eddy current probe to evaluate its feasibility for dilation measurement. It is observed that a phenomenon of a divergence of signals, akin to the lift-off point of intersection, occurs just after the transient part of the pulsed eddy current signals. Slope analysis of these diverging pulsed eddy current signals was carried out for characterisation of the probe in the air medium as well as the metallic medium present in the gap between the probe face and the test specimen.